Thermal and thermal-stress analyses of IFMIF liquid lithium target assembly

Ida, Mizuho*; Nakamura, Hiroo; Shimizu, Katsusuke*; Yamamura, Toshio*

Fusion Engineering and Design, 75-79, p.847 - 851, 2005/11

https://doi.org/10.1016/j.fusengdes.2005.06.228

no abstracts in English

None

;

PNC TN9100 96-014, 833 Pages, 1996/10

PNC-TN9100-96-014.pdf:18.58MB

None

Large-scale test on sodium leak and fire (IV); Test of sodium leak and fire using simulated piping; Run-E2

Morii, Tadashi*; *; *

PNC TN9410 87-088, 59 Pages, 1987/06

PNC-TN9410-87-088.pdf:3.33MB
PNC-TN9410-87-088TR.pdf:3.29MB

A test, Run-E2, of sodium leak from a 1/3.5 scale simulated Na piping of the secondary circuit of Monju was conducted using the SOLFA-2 in the SAPFIRE facilities. In the simulated piping, a leak hole with an area reduced by (1/3.5) from an area of 1/4Dt of the actual piping was made on the upper wall of the piping in advance. In the test, a pressure equal to the system pressure of 3.8 kg/cmg of the hot leg piping in secondary circuit of the actual plant was applied, and sodium was spilled. Spill duration was approximately 13 minutes. The test results showed that the integrity of the insulation structure around the piping will not be broken by a Na Pressure and combustion heat during an accident, therefore, the spray leakage of Na can be fully prevented. Moreover, burning rate of Na during leakage was approximately 4% of the flow rate of the Na leak. Compared to about 30% obtained from the spray combustion test conducted previously using a spray nozzle, the combustion following a realistic Na Leak from the actual piping is found to be milder than that of spray combustion.

Transient Temperature Response of In-Vessel Components Due to Pulsed Operation in Tokamak Fusion Experimental Reactor (FER)

*; Tone, Tatsuzo

JAERI-M 85-203, 35 Pages, 1985/12

JAERI-M-85-203.pdf:0.97MB

no abstracts in English

Study on High Temperature Piping(Single Wall Structure) System,A

; ; ; ; Suzuki, Katsuo

JAERI-M 8881, 73 Pages, 1980/06

JAERI-M-8881.pdf:1.8MB

no abstracts in English

Development of analysis methods for assessing the structural integrity of reactor pressure vessel

Katsuyama, Jinya; Uno, Shumpei; Li, Y.

no journal, , 

no abstracts in English